1. Field of the Invention
The present invention relates to a method of producing zirconium alloy with improved resistance to oxidation at very high temperature.
2. Description of the Related Art
Zirconium was hardly known before 1940, but gained attention mainly for its low neutron-capture cross-section and was utilized mostly in the nuclear energy-related engineering materials and nuclear energy substances. Because zirconium particularly has low neutron absorption cross-section and good resistance to corrosion, and intrinsically does not form radioactive isotopes, the material is critically used in the nuclear reactor components such as spacer grid, guide tube, heavy water reactor pressure tube, or cladding tube for a nuclear fuel rod, or alloy with uranium.Zr+2H2O→ZrO2+2H2 Oxidation of zirconium
However, zirconium alloy component generates hydrogen due to oxidation reaction between zirconium and water, which is absorbed into zirconium alloy components to form hydroxide layer(s) and, in turn, causes mechanical deformation and degradation of resistance to instability of the fuel assemblies. To overcome this shortcoming, studies have been conducted to find ways to increase resistance to oxidation and resistance to corrosiveness of the zirconium alloy. Considering the advantage of prolonged lifespan of the reactor rod structures, studies are actively conducted to develop appropriate alloys for use therein.
In the meantime, the stability of the cladding tube has been in increasing demand particularly in the event of emergency such as accident.
As learned from the incidence of reactors 1, 2 and 3 of Fukushima I Power Plant (Japan), when cooling of reactors is interrupted due to natural disasters such as earthquake or tsunami or man-made disaster, the cladding tube is exposed to high temperature so that hydrogen with high risk of detonation is massively generated due to considerably high velocity of corrosion. The hydrogen detonates when it is leaked into the containment buildings of the reactor. The hydrogen explosion in the power plant must be prevented, because this could lead to tragic disasters which could accompany leakage of radioactive substance.
Looking at the zirconium alloys currently available, these alloys do not pose considerable problems under normal condition. However, the safety of the alloys is not guaranteed when an accident occurs such as generation and detonation of hydrogen. A sufficient time has to be ensured for the management of the emergent situation before the generation of hydrogen in order to improve the safety of the nuclear power plant, and this will be possible if the nuclear fuel cladding tube has a sufficient resistance against oxidation when exposed to the emergency conditions.
The currently-available method for fabricating a zirconium alloy for use in a cladding tube basically adjusts the ratios of the alloying elements such as niobium (Nb), tin (Sn), iron (Fe), chromium (Cr), oxygen (O), or the like. However, limited oxidation resistance is expected by the method of using such alloying elements at high temperature environment. The effect of oxidation resistance that can be provided by adjusting ratios in the alloying elements is particularly insufficient to maintain resistance to oxidation under the emergency condition of the power plant which accompanies exposure to extremely high temperature for a prolonged time. The zirconium alloy has rapidly degrading oxidation resistance when temperature rises. The currently-available alloying technique, which is based on the fine adjustment of alloying composition, would not be sufficient to ensure efficiency under high temperature corrosion condition. Accordingly, it is necessary to take a step forward, for the improved accident safety of the nuclear fuel.
Meanwhile, the stability of the nuclear fuel assemblies can be increased by coating anti-oxidation material onto the surface of zirconium alloy to thus improve the resistance of zirconium alloy against oxidation at high temperature. If an anti-oxidation substance, which is stable at high temperature, is coated on the surface of the zirconium alloy to prevent oxidation from occurring when the alloy is unexpectedly exposed to high temperature environments due to changes in the environment, the oxidation reaction can be effectively restricted and less hydrogen would be generated, and therefore, danger factors such as hydrogen explosion can be prevented or reduced. However, a few substances are known for inhibiting oxidation at high temperature, and it is also a great challenge to ensure good bonding between zirconium alloy layer and coating layer of the anti-oxidation substance to prevent physical damages even at high temperature.
U.S. Pat. Nos. 5,171,520 and 5,268,946 teach a technology to coat ceramic and glass material with flam spraying to enhance wear resistance of the cladding tube.
U.S. Pat. No. 5,227,129 discloses a method for coating zirconium nitride (ZrN) with cathodic arc plasma decomposition to enhance corrosion resistance and wear characteristics.
The above patents aim to improve anti-corrosion and wear resistance of the nuclear fuel cladding tube under normal condition, and has drawbacks of limited control on the compositions of the coating layer due to use of inter-metallic compounds (ZrN, ZrC), or ceramic (zircon) or glass (CaZnB, CaMgAl, NaBSi) as the coating material. The patents also have the shortcoming of considerable differences between the coating layer and the parent material causing physical damage (e.g., crack and scrape off) due to thermal expansion and deformation. Indeed, studies reported that the layer becomes porous when oxidized at high temperature so that improvement of corrosion resistance is hardly anticipatable under emergency condition of the nuclear power plant (S. Shimada, Solid state ionics 141 (2001), 99-104; L. Krusin-Elbaum, M. Wittmer, Thin Solid Films, 107 (1983), 111-117).
Conventional studies on coatings on nuclear fuel claddings aim to overcome limited corrosion resistance by utilizing alloying elements, i.e., by forming a layer with resistance to corrosion and wear using methods such as ion implantation, Zr—N layer deposition, or the like.
U.S. Pat. No. 4,279,667 discloses a zirconium alloy structure and a processing method thereof, which use ion implantation to improve corrosion resistance.
Korean Pat. No. 2006-0022768 discloses a technology to form Zr(C, N) layer on the surface of a cladding tube by chemical vapor deposition (CVD) or physical vapor deposition (PVD) to improve corrosion characteristic of the zirconium alloy cladding tube.
However, these technologies have shortcomings in that the layer newly generated on the surface is not thick enough to effectively prevent corrosion, or due to columnar crystal structure thereof, unable to prevent oxidation due to inter-granular diffusion of oxygen. Accordingly, a process is necessary, which generates a layer that does not easily allow diffusion of oxygen on the surface of the nuclear fuel cladding tube to a sufficient thickness to prevent corrosion of the cladding tube during the normal operation condition in reactor.
In consideration of the above, the present inventors have come up with a zirconium alloy with greatly improved resistance to oxidation at very high temperature by evenly coating a pure metallic substance with excellent resistance to oxidation onto zirconium alloy with plasma spraying.